ML13196A422

NRC Staff Answer to Motion for Summary Disposition of Contention 4B - Attachment 4B-G

ML13196A422
Person / Time
Site:	Seabrook
Issue date:	07/15/2013
From:	Atomic Safety and Licensing Board Panel
To:
SECY RAS
References
50-443-LR, ASLBP 10-906-02-LR-BD01, RAS 24821
Download: ML13196A422 (4)

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NextEra Energy Seabrook, LLC (Seabrook Station, Unit 1) License Renewal Application NRC Staff Answer to Motion for Summary Disposition of Contention 4B ATTACHMENT 4B-G Purdue University researchersare working on a project to

compare the three leading se-vere accident programs, or codes, used by the nuclear power industry in the United States.

The codesÑMELCOR, MAAP4, and SCDAP/

RELAP5, all developed for different approaches and for different purposesÑhave been tested at Purdue using a hypothetical accident scenario (station blackout with no recovery of auxiliary feedwater) at a four-loop pressur-ized water reactor based on the now closed Zion nuclear power plant. Conservative analysis conditions were used to investigate the integrity of the

steam generator tubes and other components during the accidents

progression. Despite considerable differences in the codes themselves, test results show that the codes are

similar in terms of thermal-hydraulic and core degradation response.To date, plant data for an actual se-vere accident at a nuclear power plant exists only from the Three Mile

Island-2 incident in 1979. Code simu-lations for the TMI-2 scenario have

been carried out in the past with SCDAP/RELAP, MELCOR, and MAAP4, but never have the results of

the three codes for the same hypothet-ical accident been compared in detail, and never has the relative state of modeling been pursuedthis thoroughly.Karen Vierow, an assistant professor in the School of Nu-clear Engineering at Purdue University, is leading the re-search, which is being sponsored by the Nuclear Regulato-ry Commission. She has worked on the project with Yehong

Liao and Jennifer Johnson, graduate students at Purdue; Mark Kenton, a MAAP4 developer currently with Creare, Inc.; and Randy Gauntt, a MELCOR developer from Sandia National Labo-ratories. The accident simulations have been performed, and Vierow is now in

the process of analyzing the data.MELCOR was developed by Sandia National Laboratories; MAAP4 (Mod-ular Accident Analysis Program) by Fauske & Associates, Inc.; and SCDAP/RELAP5 (Severe Core Dam-age Analysis Program/Reactor Excur-sion and Leak Analysis Program) by

the Idaho National Engineering and Environmental Laboratory.Vierow talked with Rick Michal,Nuclear NewsSenior Editor, about thecode research work.Three computer programs used to simulate severeaccidents at nuclear plants are themselves

analyzed for comparison.

THE NUCLEAR NEWS INTERVIEWKaren Vierow: Severe accident code analysis Vierow: The codes have undergonesignicant upgrades over the years and are

becoming more best-estimate in nature.

March 2005NUCLEARNEWS 23Operations Interview begins on next page How did you get involved in this research?

About four years ago, I was doing some validation of the MELCOR code. I then

started studying MAAP4 and SCDAP/

RELAP5 to look into the current state of the

technology. I found that the three codes all

have their own unique features and that they

can learn from each other. That was the rea-

son I extended my work to researching all

three codes.

Who are the main users for each of the codes?MELCOR and SCDAP/RELAP5 are used by regulatory agencies and research institu-

tions to evaluate hypothetical severe accident

events, such as a station blackout or the po-

tential for a steam generator tube rupture.

MAAP4 is the severe accident code most

widely used by nuclear utilities and vendors

because of its short run time and reduced re-

quirements for code expertise. The Electric

Power Research Institute and many utilities

also use it for the NRCs Signicance Deter-

mination Process and other analyses.

In addition, MAAP4 can be used by anexisting plant to simulate how a proposed

modication would affect plant operations.

And plant designers could use any of the

codes to predict the performance of future

plants if a certain set of conditions were im-

posed on those plants.

The codes have undergone signicant up-grades over the years and are becoming

more best-estimate in nature. MELCOR

was originally intended to be a probabilis-

tic risk assessment tool; the initial objective

for the MAAP4 code was to predict severe

accidents, using simple models based on

first principles; and SCDAP/RELAP5 be-gan as a best-estimate code with physics-

based models. While still different, the ca-

pabilities and applications of the three codes have been converging over the years.

Could you provide an overview of the codes?All three codes are capable of modelingreactor coolant system response, core ma-

terial chemical reactions, core heat-up,

degradation, and relocation, heat structure

response, and other severe-accident phe-

nomena.Modeling of ssion product release and transport and containment phenomena are

integrated into the MELCOR and MAAP4

codes. SCDAP/RELAP5 is characterized

by its detailed, mechanistic models of

severe-accident phenomena; however, the

calculations can be rather time-consuming.

SCDAP/RELAP5 typically uses on the or-

der of hundreds of hydrodynamic compo-

nents to model the primary system.

MAAP4 calculations require minimal com-

putation time with simplified geometry

models. MELCOR falls in between these

two codes, being much closer to SCDAP/

RELAP5 in terms of nodal complexity. It runs at a moderately fast speed and has a large number of mechanistic models. In

early applications, MELCORs spatial dis-

cretization of a nuclear power plant con-

sisted of roughly 1030 control volumes,

and a large number of parametric calcula-

tions could be run in a short time. With the more complicated calculations that are now

being demanded of it, MELCORs run

times have been increasing and are roughly

equivalent to those of typical SCDAP/

RELAP5 calculations.

What kind of accident simulation did you run for your research?

The information input into the three codes prior to the accident simulation was made

as similar as possible, and consistent condi-

tions were placed on all of the analyses. Of

course, to arrive at a set of conditions that

the three codes could cover, various as-

sumptions had to be made that rendered the

calculation unrepresentative of plant behav-

ior after some point in the calculation. Since

the analysis scope of SCDAP/RELAP5 is

limited to the failure of the primary side

pressure boundary, the test focused on in-

vessel severe accident phenomena, with a

special interest in the steam generator tube

response to thermal transients during this

period. Its important to remember that our

objective was to compare the three codes, as

opposed to validating them.How many simulations of the same accident were performed?

I ran the analysis probably 20 times for MELCOR and SCDAP/RELAP5. Marc

Kenton performed the MAAP4 analysis

because I do not have this code at Purdue.

I found that it was difficult to reconcile all of the input for each of the codes. When I

say input, I mean the geometric description

of the plant, the initial conditions, the

boundary conditions, etc. I tried to make

them as consistent as possible for all three

codes, but sometimes it was easy to miss

something.

There are many modeling options and a lot of things that are compared. Its not just

the same geometry, but also the same

physics models that must be analyzed. So,

I have to run the codes many times to get

the parameters to be comparable.

As for computing power, Ive found that MAAP4 runs much faster than the other

two codes. With MELCOR and SCDAP/

RELAP5, it depends on what type of event

is being run. Sometimes the test will go for

a few minutes, but sometimes it will take

several hours. Its not in terms of months as

it used to be for other codes. The increase in

computer power definitely shortens our

simulation time.

Why did you pick a four-loop PWR, basedon the Zion plant, to simulate an accident?

Zion was chosen because it is a represen-tative four-loop plant in the current PWR

eet. We are also evaluating the codes for

other plant types, including boiling water

reactors and Babcock & Wilcox OTSG

(Once Through Steam Generator) PWRs.

Since Zion was one of the plants used in the famous NUREG-1150 assessment of severe

accident risks, input decks that we could start with were already available. A lot of

safety systems and normal features were disabled so that the core would melt and the

physics models in the codes would be tested 24NUCLEARNEWS March 2005 INTERVIEW

VIEROWThe onset of hydrogen production (start of fuel damage) is predicted to occur at nearly the same time by all three codes, and roughly the same amount of hydrogen is produced.

through core degradation, relocation, and many other possible serious events.

What are the results of the simulations?

The results show that the thermal-hydraulic phenomena and major in-vessel

severe accident phenomena are in good

agreement for the three codes. Also, the in-

tegral effect of diversified core models in

terms of total hydrogen production and to-

tal core debris mass slumping into the reac-tor vessels lower head are consistent for the three codes. This consistency will prob-

ably reduce the codes prediction differ-

ences for ex-vessel severe accident phe-

nomena, such as ex-vessel corium water or

corium concrete reaction, hydrogen behav-

ior in the containment, and containment

pressure response. There are also some dis-

crepancies that could be termed as minor

and that are possibly due to uncertainties in

the numerics and physics models.During the testing, several key assump-tions were made to account for known dif-

ferences in heat transfer modeling and the

representation of countercurrent natural cir-

culation of hot gases. Given these assump-

tions, the three codes predicted similar tem-

peratures in the various reactor coolant

system components. Future work could fo-

cus on resolving the modeling differences

in these few key areas.

Were there any uncertainties during the simulations?

There is regarding core degradation, inits late phase, where things are relocating

in the core. We have less detailed knowl-

edge of how the core would relocate during

a severe accident. The industry has some

good ideas, but there is minimal plant data

on it because the only actual data comes

from TMI-2, when the core partially

melted. Based on that and on experiments,

physics models were developed that were

put into the codes. Thats where some of the

uncertainties come fromÑnot having a

complete knowledge of the phenomena,

trying to use models based on smaller ex-

periments and applying the results to a full

plant.What was the most challenging task in com-paring the codes?

The hardest part was to compare the physics models. Were trying to run all

three codes on an even playing eld. There

are differences in the nature of the codes,

and each has a different philosophy because the developers of each have their own knowledge and their own way of thinking.

It was difcult to choose the right physics

models for these analyses so that each code

was given a fair chance to calculate the

same event.

What were your conclusions at the end of the research?

The key conclusion was that each of the codes has high capabilities, but that some have physics models that could be incorpo-rated into the others. So, while all the codes are impressive, each can benet from learn-

ing from the other codesÑlooking at the key

assumptions made in the other codes and see-

ing where some of the assumptions are valid

and where some models could be improved.

Could you elaborate on how and what thecodes can learn from one another?For example, SCDAP/RELAP5 has the most detailed treatment of hydrogen pro-

duction from oxidation. The hydrogen pro-

duction rate is a key indicator of the pro-

gression of a hypothetical severe accident.

Once the fuel is damaged, MELCOR has

good physics models to predict the ssion

product release and transport phenomena.

These capabilities have been removed from

recent versions of SCDAP/RELAP5, and

another code performs these calculations

for SCDAP/RELAP5 users. Perhaps the MELCOR code will incorporate some of the knowledge base from SCDAP/RELAP5

for hydrogen models or use a more paramet-

ric approach to account for uncertainties in

hydrogen-related phenomena. MAAP4, on

the other hand, is able to complete calcula-

tions in orders of magnitude less time than

the other codes due to simplied versions of

basic equations and fast-running models

based on rst principles. The assumptions

made in MAAP4 to allow it to run so fast

should be further analyzed and applied as

appropriate into the other codes.

Are there any specific improvements youcould talk about?

There were no major deciencies in the codes. Each one could, of course, add a model here or there, or make an existing model more detailed. In fact, I am working

now to develop new physics models for the

codes and modify current ones to improve

their prediction capabilities.

Could the codes be used to test reactor de-signs that may be used in a future hydrogen

economy?These three existing codes were designed to analyze light-water reactors. In a hydro-

gen economy, the reactor would be a high-

temperature design that would be cooled by

a gas or molten salt instead of water, so the

codes would have to be modified to be

valid. The geometry of the high-tempera-

ture plant is also different from that of an

LWR. The high-temperature plant could be

what is called a prismatic design, or a peb-

ble bed design that has a lot of graphite in

it. That new code would need to include

new models for the graphite behavior and

the different geometry of the core. There

would not be the traditional vertical fuel rod

assemblies that LWRs have.

These codes are big computer programs,and they have perhaps 400 000 lines of For-

tran. They have a numerical architecture.

The basic equations are all set up, and its

not an easy task to make major modifica-

tions or write a new code. Each code has a

core set of writers that can make major

modications to it.

What kind of effort would be needed to de-velop codes for Generation IV gas-cooled

reactors?

Some of this is included in my earlier statements on reactors for hydrogen pro-

duction. Gas properties would need to be

confirmed and the codes would need to be

tested for their capability to run without the light water coolant/moderator they were developed for. Changing to a gas

coolant should not be too much of a chal-

lenge because gas behavior is much eas-

ier to predict than steam/water behavior

with its phase changes between liquid and

vapor and the complicated steam/water property tables. In addition to modeling

the different fuel geometry and incorpo-

rating severe accident models for phe-

nomena peculiar to gas-cooled reactors,

events not considered for LWRs must be

modeled. These would include introduc-

tion of air or water into the primary sys-

tem that could result in graphite burning.

We believe the codes are flexible enough

that they can be modified to model Gen

IV reactors.

26NUCLEARNEWS March 2005 INTERVIEW

VIEROWEventMELCOR (s)SCDAP/RELAP5 (s)MAAP4 (s)Onset of natural circulation9 3009 0009 720Failure of surge line16 28714 95514 860 Failure of hot leg piping on pressurizer loop16 46415 72015 267 Failure of SG tubes on pressurizer loop16 55315 21014 913 TIMINGOFHEATSTRUCTUREFAILURESEventMELCOR (s)SCDAP/RELAP5 (s)MAAP4 (s)Start of core uncovery7 6807 1609615 Core completely voided11 6209 95014500 a5% cladding oxidized13 78014 80613 800 Slumping to lower head16 18916 13021 994 TIMINGOFKEYEVENTSaMAAP4 calculated a very slow rate of water level decrease at the bottom of the core, leading to a substantial delay in the voiding of the bottom node. This is, in part, due to continued slow draining of the pressurizer. (s) = second